Disused Sources Research Articles

Application of the scoping tool and AMBER model in radiological safety assessment of the barrier systems of the borehole disposal system for the disposal of disused sources in Ghana

Radiological safety assessment was carried out to assess the integrity and longevity of the barrier systems of the borehole disposal system (BDS) for the disposal of disused sources in Ghana. The design of the BDS incorporates both the engineered and natural barriers into its safety concept. The safety of the BDS requires confidence in the ability of the engineered barriers on the host environmental conditions to provide containment for the disused sources for a sufficient length of time. The results obtained from the scoping tool indicated that the dose limit was exceeded which showed that the containment provided by the engineered barriers was not sufficient to ensure safety of the disposal system for the specified environmental conditions and radionuclides inventory. Detailed modelling was performed with an AMBER model to evaluate the containment provided by the engineered barriers and the geosphere with some identified scenarios. The AMBER model results showed that the calculated peak dose from any of the disposed radionuclides for the identified scenarios was below the dose constraint of 0.3 mSv/y. The peak doses occurred around 2000–1000000 years. The peak doses were seen to be mainly derived from Ra-266 and Am-241 and/or their daughters. The disused sources could be disposed of safely in the disposal system either in an oxidizing or reducing fractured/porous flow environment without posing any significant radiological threat to humans and the environment.

Monte Carlo optimum management of 241Am/Be disused sealed radioactive sources

The optimum encapsulation of 241Am/Be disused sealed radioactive sources (DSRS) based on PHITS Monte Carlo simulations for their long-term storage in Cameroon was performed. The country capacity for the management of disused neutron sources was also evaluated and showed that a Am1 P60 capsule is sufficient for the total available inventoried 241Am/Be DSRSs. The effective dose rate was computed in the enclosures of the DSRS container, which will be temporarily stored in the centralized radioactive waste facility. The obtained results were in agreement with the ALARA principle for the exposure rate optimization and the obtained exposure dose rates were found to be 1.830 μSv/h (horizontal calculation) and 0.137 μSv/h (vertical computation) which values are lower than the 2.5 μSv/h acceptable limit for the public area. The dose profile for 241Am/Be source obtained, the neutron flux, and gamma generated from neutron absorption showed agreement with the research hypothesis. The Monte Carlo assessment achieved in the present research will be useful for dismantling and preparing the waste package for long-term storage.

Development and application of an approach for safety assessment of radioactive waste storage facilities for accidental scenarios

Worldwide there is a huge amount of radioactive waste, including disused sources, decommissioning waste, and naturally occurring radioactive material (NORM) waiting for final disposal, the so-called storage facilities. Results of safety assessment of such facilities are usually required in decision-making during design, modification, safety improvements, periodic safety reviews, and licensing activities. Quantitative safety assessment methodologies used in many areas of nuclear industry involves the evaluation of risk through the definition of scenarios, likelihoods, and consequences, for normal operation and accidents. There are many techniques that can be used in each one of these steps. For screening the accident scenarios, qualitative techniques such as failure modes and effect analysis (FMEA) are available. For a quantitative assessment of occurrence probabilities of undesired events, logical and graphical tools such as fault tree analysis (FTA) are employed. Consequence assessments involve the dose assessment in the workers and the impact of the released materials in the environment and of public exposure to radiation. This work analyzes the application of these traditional safety assessment approaches to storage facilities and how they can be applied to complement specific methodologies used in this area, such as the Safety Assessment Driving Radioactive Waste Management Solutions (SADRWMS), implemented in Safety Assessment Framework (SAFRAN) software tool, made available by International Atomic Energy Agency (IAEA).

End of Life Decisions for Sealed Radioactive Sources

Sealed radioactive sources are encountered in a wide variety of settings-from household smoke detectors and instrument check sources through fixed industrial gauges, industrial radiography, and well logging sources, to irradiators and medical teletherapy devices. In general, the higher the level of activity in the sealed source, the stricter the regulatory control that is applied to its use, control, and ultimate disposition. Lower levels of attention and oversight can and do lead to sources ending up in the wrong place--as orphan sources in uncontrolled storage, disposed in a sanitary landfill, melted down in metal recycling operations and incorporated into consumer products, or handled by an unsuspecting member of the public. There is a range of issues that contribute to the problem of improper disposal of sealed sources and, in particular, to disused source disposal. Generally licensed sources and devices are particularly at risk of being disposed incorrectly. Higher activity generally licensed sources, although required to be registered with the (NRC) or an Agreement State, receive limited regulatory oversight and are not tracked on a national scale. Users frequently do not consider the full life-cycle costs when procuring sources or devices and discover that they cannot afford and/or are unwilling to pay the associated costs to package, transport and dispose of their sources properly. The NRC requirements for decommissioning funding plans and financial assurance are not adequate to cover sealed source transport and disposal costs fully. While there are regulatory limits for storage of disused sources, enforcement is limited, and there are only limited financial incentives in a small number of states for owners to dispose of the sources. In some cases, the lack of availability of approved Type B shipping casks presents an additional barrier to sealed source disposal. The report of the Disused Sources Working Group does an excellent job of framing these issues (www.disusedsources.org/wp-content/uploads/2014/12/DSWG-Report-March-2014.pdf). This article reviews both the issues and the report's recommendations, which are designed to improve sealed source control and encourage proper disposal of disused sources.

Recycling of a Disused 137Cs Sealed source

Recycling of disused sealed radioactive sources (DSRS) is one of the milestones of radioactive waste management strategy that prompts the minimisation of the radiological hazard to human health and the environment. In this respect this study spotlights recovering and purification of 137Cs as a solution (with a radioactivity of 191.5 ± 3.9 MBq) from an old disused source. The method was designed to fulfil the regulations and legislations of radiation protection giving exposure doses of the operating staff within the international permissible limits. Separation of the chemical impurities was performed by hydroxide precipitation based on pH values and solubility constants. X-ray fluorescence analysis illustrated a removal efficiency of 99% for iron and chromium, whereas the concentration of nickel and manganese decreased to 80 and 7 mg L-1, respectively, in the purified 137Cs solution. The radiochemical recovery of the purification process was determined using gamma-spectroscopy and found to be 94.4%. The purified 137Cs solution was then used to prepare a gel-source with 20 mL cylindrical geometry. Calibration was carried out at a secondary standard laboratory “National Radioactive Measurement Laboratory” (NRML) that is traceable to the primary standard “National Physics Laboratory” in the UK (NPL), giving an activity of 10.59 MBq with relative uncertainty of 0.4-0.5%. The preparation procedure could be considered as an original, facile, feasible and cost effective method that implies beneficial reuse of a DSRS in terms of volume minimization of radioactive wastes.

Removal of Alcyon II, CGR, MeV <sup>60</sup>Co Teletherapy Head and Evaluation of Exposure Dose

Teletherapy units containing 60Co source are used in Egypt and worldwide for treatment of cancer diseases. One of the widely used in Egypt is Alcyon CGR MeV. Different radiotherapy facilities need to unload this 60Co unit after being disused. The head containing the source was removed and put into approved package for transportation, after the source kept in safe position. This study describes the planning and removing process of the disused sources and, transports it into its head. The personal doses measured during the process. This work pre-estimates the doses received to workers during the removing process. The workers received doses of approximately 55 μSv during the whole process. This study shows that the original lead head can be used as the source container of this 60Co unit but not as a long term storage option. The 60Co machine was smoothly dismantled and transported by trained workers and driver, using planned and controlled radiation protection, followed by ALARA (as low as reasonably achievable) rule.

Lack of international consensus on the disposition and storage of disused sealed sources

A lower-activity analogue of the trans-national problem of spent fuel management and disposal is the global problem of radioactive sealed source [source: The IAEA definition of a sealed source is “Radioactive material that is permanently sealed in a capsule or closely bonded and in a solid form.” Taken from glossary of Nuclear Waste Data Management found at http://www-ewmdb.iaea.org/showhelp.asp?Topic=8-1-1.] disposal. Sources are found in almost every country in the world because of their beneficial medical and commercial or industrial applications. Some of the isotopes used have short half-lives—iridium-192 (Ir-192), 73.8 days—while others have very long half-lives—americium-241 (Am-241), 432 years or plutonium-239 (Pu-239), 24,130 years. It is critically important, particularly for longer-lived isotopes, to find final disposition pathways. Lack of a permanent disposition pathway such as recycling or irretrievable disposal creates numerous problems, including the potential loss of regulatory control, which increases the risk of inadvertent or deliberate misuse of the material. The misuse of radioactive materials has the potential for substantial public health and economic damage. Disused sources also pose an inherent risk to the end-users from a liability, safety, and public health perspectives. This paper examines various disposition pathways employed by several key source manufacturing or possessing nation-states for disused sources. Examples of source disposition pathways include long-term storage, deep geological disposal, borehole disposal and shallow land burial. The Off-Site Source Recovery Project (OSRP), part of the office of Global Threat Reduction Initiative (GTRI), acts as an intermediary in the recovery and ultimate disposition of US origin sealed radiological materials. Several concepts that could help mitigate the challenge of a lack of long-term disposition options for sources are available, but these tools have not yet been applied by most nation-states. For example, regional consolidation and repatriation of sources to the country of manufacture would ease or eliminate the need for in situ disposal or storage in a number of developing nation-states.

Nuclide Distribution between Steelmaking Phases upon Melting of Sealed Radioactive Sources Hidden in Scrap

For sealed radiation sources widely used in industry, medicine and research, many states have clear regulations concerning the management of disused sources. Once and again, however, disused sources get lost to regulatory control and may eventually be discarded as plain steel scrap. The case that such lost sources pass the steelmakers’ gate monitors and end up in the meltshops cannot totally be disregarded. This paper lists the relevant nuclides presently or formerly in use as radiation sources and reports on their expected distribution between steel melt, slag, dust and off-gas upon accidental meltdown. Equilibrium distributions based on thermodynamic calculations are presented for Co, Ni, Sr, Cs, La (substituting Pm), and U (substituting Ra, Am, Pu and Cm). In addition, realistic distribution ratios are given taking into account effects of incomplete metal/slag separation and carry-over of fine droplets from the melt to the dust due to bubble bursting. Three groups of nuclides are identified, namely (a) strong γ-emitters (Co 60, Kr 85, Cs 137, Ir 192), (b) weak γ-emitters (Ra 226, Am 241), and (c) β- and α-emitters (Ni 63, Sr 90, Pm 147, Pu 238/239, and Cm 244). The strong γ-emitting nuclides can be, and are increasingly, monitored on-line in the steelworks. For the other groups, an on-line detection is technically not possible at the present time. Global regulations and control managements must enforce the safe disposal of sources to make sure that disused sources cannot find their way into steel melt-shops anymore.